Workshop on Operational and Safety Aspects of
Sodium Cooled Fast Reactors,
Vienna, Austria, 23-25 June 2010
SODIUM COOLED
FAST REACTORS:
VITAL SAFETY ISSUES AND
APPROACHES TO THEIR RESOLUTION
V. Rachkov
Institute for Physics and Power Engineering
(IPPE), Obninsk, Russia
OUTLINE
1. Background
2. Tasks and approaches to safety
analysis
3. Proposals on cooperation
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1. Background
Studies on sodium cooled fast reactors (SFR) have been
carried out in the USSR and in Russia for over half a
century. During this period the following reactors have
been designed, constructed and put into operation:
• BR-5/10 research fast reactor, which was in operation at the
IPPE, Obninsk from 1959 to 2002;
• BOR-60 experimental fast reactor, which has been in operation
at the RIAR, Dimitrovgrad from 1969 till now;
• BN-350 prototype reactor, which was in operation in
Shevchenko/Aktau from 1972 to 1999;
• Commercial BN-600 reactor having been operated at Beloyarsk
NPP, Zarechny over 30 years from 1980 till now;
• Commercial BN-800 reactor is now under construction on
Beloyarsk NPP site (its start-up is planned for 2014).
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Sodium leaks in fast reactors
Traditionally, when considering SFR
safety, much attention is paid to the
issues concerning high chemical
activity of sodium. In Russia we have
gained considerable experience on
sodium leaks from the circuits and
leaks in the steam generators (SG).
Number of leaks
from the circuits
Number of SG
leaks
19
-
BOR-60
-
1
BN-350
15
14
BN-600
27
12
Total
61
27
Reactors
BR-5/BR-10
All sodium leaks in BN-600 took place in the early stage of reactor
operation and were timely detected and confined by the relevant
safety systems and personnel.
The last sodium leak from the circuit occurred more than 16 years
ago. During over 19 years steam generators operated without any
inter-circuit leaks in spite of multiple replacements of SG modules
carried out in this time period.
None of occurred sodium leaks caused exceeding of safety limits
radiation effect on the inhabitants and the environment.
4
Load factor of the BN-600 reactor power unit for
the whole period of its operation
100
90
83,53
80,29
80
74,11
71,76 72,75 72,48 73,46
КИУМ,%
70
76,6 75,89
69,83
79,89
78,19
76,43
76,32
70,31
73,23
72,97
77,35 75,74
80,04
77,75 78,6 77,78 77,49 76,53
65,91
60 56,51
50
47,93
40
30
20
10
0
1982 1983 1984 1985 1986 1987 1988 1989 1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004 2005 2006 2007 2008 2009
Sodium fast reactor operating experience gained in Russia
makes it possible to not only draw a conclusion about
commercial development of SFR technology, but also
demonstrates the possibility of reliable and safe operation
of such reactors.
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From BN-600 to BN-800
The design of BN-800 reactor which is under
construction
is a logical advancement of the
BN-600 design with enhanced safety features
meeting modern requirements, such as:
• application of decay heat removal system
independent from the third circuit;
• presence of additional passive reactor safety
system using safety rods suspended in sodium
flow;
• arrangement of core debris catcher in the lower
section of the reactor vessel.
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Future plans
This January Federal Target-oriented Program (FTP)
“The New Generation Nuclear Power Technologies
for 2010-2015 Period and with Outlook to 2020” was
approved by the RF Government.
Within the framework of this FTP advanced SFR of the
next generation (BN-1200) with significantly
improved technical and economical characteristics
will be designed.
BN-1200 would be basis in achieving the goal of
commercialization of SFR and closed nuclear fuel
cycle.
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2. Safety approaches for new designs
The main feature of the next stage of SFR
development is simultaneous achievement of the
goals of improvement of economical and safety
characteristics.
In our vision the main path for this is max use of
inherent safety characteristics, as well as safety
systems based on passive operating principles.
This approach was used in optimization of
characteristics of BN-1200 in the stage of its
conceptual design. The main characteristics of the
BN-1200 reactor taken for design development are
presented in the following table.
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The main characteristics of the BN-1200
Parameters
Values
Reactor thermal power, MW
2900
Reactor electric power, MW
1220
NPP efficiency, %
− gross
− net
Number of heat removal loops
42.0
39.0
4
Reactor lifetime, years
60
Primary sodium temperature (IHX outlet/inlet), 0С
410/550
Secondary coolant temperature (SG outlet/inlet), 0С
355/527
Parameters of the third circuit:
− pressure (at the SG outlet), MPa
− temperature (at the SG outlet), 0С
− feed water temperature, 0С
− reheated steam temperature in the steam-steam
reheater, 0С
14.0
510
240
250
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BN-1200 - new decisions
Main decisions adopted in the BN-1200 reactor design in order
to improve its economic and safety features are as follows:
• integral arrangement of the primary circuit in the reactor
vessel this practically eliminating the possibility of
radioactive sodium release from the primary circuit;
• simplification of the reactor refueling system owing to
elimination of the intermediate storage drums for fuel
subassemblies and arrangement of high-capacity in-vessel
storage (IVS);
• change-over from section-module steam generator design to
integral SG design;
• use of passive reactor safety systems based on diverse
operating principles as well as of decay heat removal system
independent from the main heat removal loops.
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Closed fuel cycle safety
Change-over to closed fuel cycle of nuclear power is critical
and conceptual choice of Russia. Within the framework of
FTP it is planned:
• creation of semi-industrial plant for BN MOX fuel fabrication;
• development and tests of high-density fuels;
• development of the new generation structural materials;
• development of dry methods of reprocessing of BN SNF;
• experimental studies on radwaste final disposal.
In this stage, the key safety aspects are as follows:
• assurance of nuclear and radiation safety in all CNFC
stages;
• overall decrease of energy system environmental impact.
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Safety approaches in the area of fuel cycle
Some thoughts.
There is no doubt in order to get public support it is desirable to
eliminate long live radwaste toxicity in repository (for example through
transmutation of all minor actinides and long lived fission products),
however we should not forget that the main our goal is nuclear power
competitiveness (with all the additional transmutation costs). This lead
us to the idea of the need for finding of optimum transmutation level.
This optimization have to be done by taking into account that assurance
of radiation safety associated with on ground fuel cycle facilities are
not less, but might be even more important.
Obviously, it is interested to consider the possibility and expedience of
unification of the national and international standards on releases
concerning reactor plants and nuclear fuel cycle facilities.
In particular, defining requirements to the possibility and extent of
discharge of liquid radioactive waste of various categories is a relevant
objective.
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Development of research infrastructure
In the institutes and enterprises of the Rosatom there
is an extensive research and trial infrastructure,
which has been and can be further used for carrying
out necessary R&D work on SFR safety analysis, as
well as verification of computer codes used for these
purposes.
Irradiation of experimental devices with various fuel
compositions and structural materials can be also
performed in BOR-60 experimental reactor, as well as
in the BN-600 reactor.
It is planned to use in the future MBIR experimental
reactor for this purpose.
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Experimental facilities of the IPPE
IPPE possesses several dozens of experimental
facilities related to various aspects of sodium
cooled fast reactors. All these facilities can be
divided into the categories according to their
purpose:
•SFR core neutronics;
•studies of fuel and materials under radiation;
•SFR safety;
•thermal hydraulics;
•sodium coolant technology.
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Experimental facilities of the IPPE for
neutronics and safety studies
• BFS-1 critical facility (studies on neutronics of the core, blankets, in-vessel fuel
storage and in-vessel shielding on the full-scale models of research and power
fast reactors up to 1000 MWth);
• BFS-2 critical facility (studies on full-scale models of the core, blankets, invessel fuel storage and in-vessel shielding).
• PLUTON test facility (studies on the processes of materials movement caused
by interaction of uranium-containing corium simulators with sodium);
• AR-1 test facility (studies on thermal-physics processes; studies of flow
stability and heat transfer characteristics in case of coolant boiling);
• Integral model of fast reactor, facility for testing decay heat removal system;
• IK-MT liquid metal test facility (carrying out tests of elements of automatic
safety systems of sodium-water steam generators for SFR and studies on the
small leaks in sodium-water steam generators);
• SAZ facility for testing safety system of NPP with SFR).
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Experimental facilities of the Rosatom
institutions for safety studies
RIAR - BOR-60 experimental reactor;
OKBM - water facility for full-scale tests of the
main pumps for SFR;test facility for studies
on the core debris catcher for the case of
beyond design accident in SFR with core
meltdown;
OKB Gidropress - complex of test facilities
designed for carrying out studies on thermal
hydraulics of sodium-water steam generators
and structural and vibration tests of the main
elements of the steam generators.
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Experimental facilities for new designs
FTP implies creation of the new experimental
facilities and equipment, upgrading and
development of existing test facilities required for
carrying out R&D work for justification of SFR.
The R&D work programs include upgrading of the
complex of the large critical facilities BFS and
designing and construction of the multi-purpose
research fast reactor MBIR with sodium coolant,
which is planned to use for tests of the fuel and
structural materials.
Now the preparation work is carried out on terms of
reference for designing research reactor MBIR.
Start-up of MBIR reactor is planned for 2019.
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The main parameters of MBIR reactor
Parameters
Values
Thermal power, MW
~ 150
Electric power, MW
~ 40
2
~ 6.0·1015
Max neutron flux, n/cm ·s
Standard fuel
Vipac-MOX, (PuN+UN)
Experimental fuel
Innovative types of fuel,
fuel with MA
Core height, mm
600
Max power density, kW/L
1100
Max annual fluence, n/cm
2
Lifetime, years
Number of independent loops with various coolants
Total number of experimental subassemblies and irradiation devices for
radioisotopes production
Number of experimental channels
Number of horizontal experimental channels
Number of vertical experimental channels
~ 1·1023 (up to 45 dpa)
50
Up to 4
Up to 12 (core)
Up to 5 (radial shielding)
Up to 3 (core)
Up to 6
(outside reactor vessel)
Up to 8
(outside reactor vessel)
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MBIR design and experimental capabilities
Experimental devices:
- 1 test loop in the core central area
D~130 mm, Fn~5·1015 n/(cm2·s)
- 2 test loops in the reflector
D~130 mm, Fn~1.8·1015 n/(cm2·s)
Loop channels coolants:
- liquid metals
- gas
- molten salts
Vertical experimental channels (VEC):
3 channels in the core
D~60 mm, Фn~(3.2-5)·1015 1/(cm2·s)
Material test subassemblies:
up to 15 channels in the core and reflector
D~60 mm, Фn ~ 5·1015 1/(cm2·s)
Horizontal experimental channels (HEC):
up to 6 channels outside reactor vessel
D~150-200 mm, Фn ~ 0.5·1014 1/(cm2·s)
Inclined experimental channels (IEC):
up to 7 channels outside reactor vessel
D~150-300 mm, Фn ~ 0.5·1014 1/(cm2·s)
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Development of computer codes
For the purpose of safety analysis over 20 computer codes are
used in Russia allowing to simulate the whole range of
transients, abnormal operating conditions and design basis
and beyond design accidents typical for SFR.
According to FTP, the new generation integrated code systems
will be designed for the analytical studies on safety of
advanced NPP with nuclear fuel cycle, including BN-1200.
There is a goal to make certification of all computer codes
used for analysis of safety and other characteristics of
advanced SFR. Besides, FTP implies upgrading of existing
codes, including development on their basis of the new
generation integrated code systems capable of designing
reactor components and systems of advanced NPP with SFR,
making their comprehensive safety analysis, optimization of
NPP characteristics and information support of NPP
operation.
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3. Potential cooperation areas
The following promising topics can be proposed for scientific and
technical cooperation in the area of SFR operation and safety:
• specification of safety criteria and requirements to new generation
of fast reactors;
• upgrading of the computer codes and their verification by means of
carrying out benchmarks. (Benchmarks on the analysis of severe beyond
design accidents are of special interest, in particular, those on the accidents
with the core meltdown, as well as on estimation of degree of influence of
some parameters on the course of certain beyond design accident and its
consequences, for instance, reactivity feedback coefficients);
• exchange of available experimental data for verification of computer
codes;
• use of the available experimental facilities for arranging and
carrying out experiments aimed at obtaining of necessary data for
verification of computer codes;
• information exchange on current status of activities in SFR area
including operation and safety issues.
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Proposals for joint studies
• Joint activities on development of MBIR design, construction and
implementation of experimental studies
• inherent safety characteristics of advanced SFR and their
contribution to reactor safety assurance;
• role of active and passive safety systems in their optimal
combination from the standpoint of safety and technical and
economical characteristics;
• various approaches to the mitigation of consequences of severe
beyond design accidents and their effectiveness;
• effect of various approaches to safety assurance on the economical
characteristics of SFR, including evaluation of potentiality of
decreasing SFR cost by means of application of passive safety
systems;
• effect of the basic conceptual approaches (rated power level, NPP
configuration, concepts of the main components and systems of
the NPP, etc.) on SFR economical characteristics, etc.
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